1. Field of the Invention
This invention relates to an apparatus and method for simultaneously loading a reinforcing sleeve and expansion mandrel into a heat exchange tube in order to expedite the installation of the reinforcing sleeve within the tube.
2. Description of the Prior Art
Devices for positioning reinforcing sleeves within the heat exchange tubes of steam generators are known in the prior art. One such device developed by the personnel of the Westinghouse Electric Corporation is capable of feeding such reinforcing sleeves through the open ends of the tubesheet of a nuclear steam generator by means of a pair of hydraulic grippers, each of which includes an expandable bladder. The frame of this device is light enough to allow it to be manipulated within the radioactive, primary side of a nuclear steam generator by a remotely operable service arm (or ROSA) that has also been invented and developed by personnel of the Westinghouse Electric Corporation. Once situated, the sleeve-feeding mechanism of this device is locked into position adjacent to the sleeve to be repaired by one or more hydraulically operated cam-lock devices that are insertable within and expandable against the open ends of tubes that are not being repaired. After the device has been secured into a proper position, the hydraulic grippers are actuated. The feeding of the sleeve is accomplished by alternatively actuating and deactuating the hydraulic grippers, and by reciprocating one of the grippers relative to the other. A complete description of this sleeve loading mechanism is set forth in U.S. patent application Ser. No. 785,291 filed Oct. 3, 1985, by Thomas E. Arzenti and William E. Pirl, now U.S. Pat No. 4,711,526, and a complete description of a hydraulically operated camlock device for mounting such tools within such steam generators is described in U.S. patent application Ser. No. 832,940 filed Feb. 26, 1986 by David J. Fink. Both of these applications are assigned to the Westinghouse Electric Corporation, and both are completely incorporated herein by reference.
Generally speaking, the ultimate purpose of such maintenance devices is to prevent radioactive water from the primary side of the generator from breaking through the weakened walls of corroded heat exchange tubes. If such a break should occur, radioactive water from the primary side would contaminate the non-radioactive water present in the secondary side of the generator. Since the water in the secondary side is used to form the steam which ultimately drives the electric turbines in the plant, it is important that this water remain uncontaminated. However, before the utility of such devices can be specifically appreciated, a more detailed understanding of the structure and maintenance of nuclear steam generators is necessary.
Nuclear steam generators are comprised of three principal parts, including a primary side, a tubesheet in which a bundle of U-shaped tubes are mounted, and a secondary side. The tubesheet and U-shaped tubes hydraulically isolate the primary and the secondary sides of the steam generator while thermally connecting them together, so that heat from the radioactive water in the primary side is transferred to the non-radioactive water in the secondary side. This heat transfer is accomplished by the U-shaped tubes mounted in the tubesheet that extend throughout the secondary side of the steam generator. The inlet and outlet ends of these U-shaped tubes are mounted in the side of the tubesheet that faces the primary side of the generator. The primary side in turn includes a divider plate that hydraulically isolates the inlet ends of the U-shaped tubes from the outlet ends. Hot, radioactive water heated by the nuclear reactor is admitted into the section of the primary side containing all the inlet ends of the U-shaped tubes. This hot water flows through these inlets, up through the tubesheet, and circulates around the U-shaped tubes that extend within the secondary side of the steam generator. This hot, radioactive water transfers its heat through the walls of the U-shaped tubes to the non-radioactive water that surrounds the tubes in the secondary side of the generator, thereby converting this feed water into non-radioactive steam. After the nuclear-heated water circulates through the U-shaped tubes, it flows back through the tubesheet, through the outlets of the U-shaped tubes, where it is ultimately circulated back to the nuclear reactor. The use of such a heat exchanger as a hydraulic interface between the nuclear core and the steam used to turn the generators advantageously confines the radioactivity generated by the core to a relatively small region of the plant.
Over long periods of time, the heat exchanger tubes of such nuclear steam generators can suffer a number of different types of corrosion degradation, including intragranular stress corrosion cracking. In situ examination of the tubes within these generators has revealed that most of this intragranular stress corrosion cracking occurs around the tubesheet region of the generator, where the inlet and outlet ends of the U-shaped tubes extend through the bores in the tubesheet. Often there is some annular space between the outer walls of the tubes and the walls of the tube-receiving bores in the tubesheet. Experience has shown that potentially corrosive sludges can accumulate on the upper surface of the tubesheet and flow down into these annular spaces over long periods of time. To prevent these potentially corrosive sludges from collecting within these annular spaces from the effect of gravity, the heat exchange tubes are often radially expanded by means of a mechanical or hydraulic mandrel to minimize the clearance between the outer walls of the tubes and the inner walls of the bores in the tubesheet through which they extend. However, some of these potentially corrosive sludges can still collect in the very small annular spaces between the tubes and the bores of the tubesheets that are left after the tubes are expanded. Moreover, the relatively poor hydraulic circulation of the water in these regions tends to maintain the sludge in these spaces and to create localized "hot spots" in the tubes adjacent the sludge. The heat radiating from these "hot spots" may assist in the corrosion processes that operate on the exterior surfaces of the heat exchange tubes in chemical combination with the corrosive species in the sludge. While most nuclear steam generators include blow-down systems for periodically sweeping the sludge out of the generator vessel, the sludges in the annular crevice regions are not easily swept away. Despite the fact that the heat exchange tubes of such nuclear steam generators are typically formed from corrosion-resistant Inconel.RTM., the constant exposure to corrosive sludges and heat, in combination with the mechanical stresses induced in these walls as a result of the hydraulic or mechanical expansion, can ultimately cause the heat-exchange tubes to corrode and crack due to intragranular stress corrosion. This, in turn, can allow radioactive water from the primary side of the steam generator to leak into the secondary side, thereby radioactively contaminating the steam produced by the generator.
Such radioactive contamination of the generator steam can be avoided if certain maintenance procedures, such as tube sleeving, are undertaken before the walls of the tubes crack. In such sleeving operations, a reinforcing sleeve is slid up the heat exchanger tubes in the sections of the tubes surrounded by the tubesheet, and then hydraulically expanded and rolled into the inner walls of the tubes. The end result is that the sleeve forms a fluid "bridge" across the weakened walls of the repaired heat exchange tube.
Such sleeving operations have proven to be very effective in extending the useful lifetime of the nuclear steam generator. Unfortunately, they are also quite expensive since the steam generator has to be completely shut down and taken off-line. Such downtime can cost the utility involved over $500,000 per day. It is therefore desirable that such sleeving operations be accomplished as rapidly as possible. While the sleeve loading tool described and claimed in the previously mentioned U.S. patent application Ser. No. 785,291 is among the fastest and most effective tools known for positioning reinforcing sleeves across corroded sections of heat exchanger tubes, it has several design limitations that prevent it from positioning as many sleeves as possible within a given unit of time. For example, the entire tool must be robotically re-positioned in the tubesheet whenever a new tube is to be sleeved. Additionally, as this tool is only capable of positioning a sleeve and not a hydraulic expansion mandrel, the maintenance operator must separately position the mandrel within the sleeve after the sleeve has been properly positioned. The separate positioning of first the sleeve and then the mandrel is time consuming, and causes the maintenance operator to be exposed to a significant dose of radiation.
Clearly, there is a need for improved sleeving devices capable of installing reinforcing sleeves in heat exchanger tubes in shorter time periods. Ideally, such a device should be easily and remotely manipulable within the radioactive environment of the primary side of the generator by means of commercially available robotic arms. Finally, it would be desirable if the operation of the device somehow combined the sequential steps of first positioning the sleeve and then positioning the mandrel so that the entire operation was substantially accelerated, and the maintenance operator's exposure to potentially harmful radiation was minimized.